1. Field of the Invention
This invention relates to an improved heater assembly for model steam generators which accurately simulates the heat flux patterns of the heat exchange tubes in nuclear steam generators by means of a relatively simple and inexpensive boiling-condensing thermosyphon.
2. Description of the Prior Art
Model steam generators for monitoring the amount of corrosion degradation occurring within the heat exchange tubes of a nuclear steam generator are known in the prior art. Generally speaking, such model generators operate by subjecting an array of sample heat exchange tubes to the same heat, pressure and chemical conditions which surround the heat exchange tubes in nuclear steam generators. If these conditions are accurately simulated, the amount of corrosion which occurs in the sample tubes of the model steam generator will provide an accurate indication of the tube corrosion present in the nuclear steam generator being monitored. Such model steam generators are a particularly useful form of corrosion monitor, because they obviate the need for shutting down the nuclear plant and sending technicians into the radioactive interiors of the generators. However, such model steam generators are useful only insofar as they are capable of accurately simulating the heat, pressure and chemical conditions which exist inside the nuclear plant. Any material departures from these conditions will adversely affect the accuracy of the model steam generator.
In order to understand the difficulties in building a practical model steam generator which provides an accurate monitor for heat exchange tube corrosion, one must first understand how nuclear steam generators are generally constructed, and what chemical and hydraulic conditions are responsible for tube corrosion.
Nuclear steam generators are comprised of three principal parts, including a secondary side and a tubesheet, as well as a primary side which circulates water heated from a nuclear reactor. The secondary side of the generator includes a plurality of U-shaped tubes, as well as an inlet for admitting a flow of feedwater. The inlet and outlet ends of the U-shaped tubes within the secondary side are mounted in the tubesheet which hydraulically separates the primary side of the generator from the secondary side. The primary side in turn includes a divider sheet which hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot water flowing from the nuclear reactor is admitted into the section of the primary side containing all of the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates around the U-shaped tubes which extend within the secondary side of the steam generator. The heated water transfers its heat through the walls of the U-shaped tubes to the feedwater flowing through the secondary side of the generator, thereby converting the feedwater to steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is recirculated back to the nuclear reactor. The inlet ends of the U-shaped tubes are known as the "hot legs", and the outlet ends of these tubes are known as the "cold legs".
The heat exchange tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including denting, stress corrosion cracking, intragranullar attack, and pitting. In situ examination of the tubes within these generators have revealed that most of this corrosion degradation occurs in what are known as the crevice regions of the generator. Such crevice regions include the annular space between the heat exchange tubes and the tubesheet, as well as the annular clearance between these tubes and the various support plates in the secondary side which are used to uniformly space and align these tubes. Corrosive sludge tends to collect within these crevices from the effects of gravity. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these crevices, and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" acts as a powerful catalyst in causing the exterior surface of the heat exchange tubes to chemically combine with the corrosive chemicals in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the crevice regions are not easily swept away by the hydraulic currents induced by such systems. Despite the fact that the heat exchange tubes of such nuclear generators are typically formed from corrosion-resistant Inconel stainless steel, the combination of the localized regions of heat and corrosive sludges can ultimately cause the heat exchange tubes to crack, and leak radioactive water from the primary side into the secondary side of the generator. However, this need not occur if the heat exchange tubes are provided with internally reinforcing sleeves before the corrosion causes cracks in the tube walls.
Model steam generators were developed in order to accurately monitor the amount of corrosion degradation occurring in the heat exchange tubes of a particular nuclear steam generator, in order that these tubes might be sleeved before any of the tube walls crack. Such model steam generators have been found to be a particularly accurate way of ascertaining the amount of corrosion degradation occurring in the heat exchange tubes of a nuclear steam generator, because the particular amount of corrosion which the feedwater chemistry and thermohydraulics of the particular generator will induce in a particular set of tubes is virtually impossible to predict by purely theoretical models.
Unfortunately, prior art model steam generators are not without significant shortcomings. For example, many of these model generators utilize a pump driven, forced circulation of water through the primary side of the generator. While such a system is capable of simulating the interior temperature and pressure conditions of the heat exchange tubes used in the nuclear steam generator being monitored, the use of a pump in the primary side renders the model as a whole bulky, expensive and maintenance-intensive.
To obviate the problems associated with such pump-driven, forced-circulation primary systems, some prior art systems have attempted to use a circulating "reflux mode" type of primary system wherein steam is injected into sample heat exchange tubes which are plugged at the ends which extend into the secondary side of the generator. As heat is transferred from the steam to the feedwater circulating through the secondary side of the generator, the steam condenses into liquid water which runs down the inner walls of the sample tubes. However, while this particular system obviates the need for an expensive and bulky pump, such "reflux mode" primary systems create other problems that are as yet unsolved. The first of these problems is that the sample tube which extends into the primary side of the model steam generator may be no more than about six inches long, or the pattern of heat flux throughout the walls of the tube will become so non-uniform that the model generator will not be accurately simulating the heat flux patterns of the tubes in the nuclear steam generator. Moreover, even though the use of relatively short sample tubes reduces many of the accuracy-destroying heat flux irregularities, it does not eliminate these irregularities altogether; a significant amount of such heat flux non-uniformities remain. Finally, the use of such short tubes in the model steam generator makes it difficult, if not impossible, to accurately simulate the thermohydraulic circulation patterns around the support plate regions of the secondary side of the nuclear steam generator being monitored. This further reduces the ability of model steam generators using such "reflux mode" primary systems to accurately simulate the conditions within a nuclear steam generator.
Clearly, a need exists for a heating system in a model steam generator which obviates the need for an expensive and bulky circulating pump, but which is fully capable of producing a heat flux pattern within the sample heat exchange tubes which accurately duplicates the heat flux patterns of the tubes used in the nuclear steam generator being monitored. Ideally, such a primary heating system should further allow the use of sample tubes which extend into the secondary side of the model to an extent sufficient to duplicate the thermohydraulic circulation patterns which exist around the support plates in the generator. Finally, it would be desirable if such a primary heating system required little or no maintenance, and was simple in construction, reliable, energy-efficient, and small enough in size to be easily installed into an existing plant facility.